High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka

Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01

https://doi.org/10.1016/j.anucene.2023.110144

Study on safety characteristics of a sodium-cooled fast reactor with negative void reactivity during initiating phase in severe accident

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

https://doi.org/10.1080/00223131.2019.1709996

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

https://doi.org/10.1016/j.anucene.2019.107188

Behaviors of high-burnup LWR fuels with improved materials under design-basis accident conditions

Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

A Study on self-terminating behavior of sodium-concrete reaction

Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

Journal of Nuclear Science and Technology, 53(12), p.2098 - 2107, 2016/12

https://doi.org/10.1080/00223131.2016.1199979

A sodium concrete reaction (SCR) is one of the important phenomena to cause the structural concrete ablation and the release of H gas in the case of sever accident of sodium cooled fast reactors. In this study, the long-time SCR test had been carried out to investigate the self-termination mechanism. The results showed the SCR terminated even if the enough amount of Na remained on the concrete. The quantitative data were collected on the SCR terminating such as temperature and H generation. The reaction products, which became the small solids in liquid Na were transferred with slurry state by generated H bubbles. Though the Na transfers actively and ablated the concrete surface with the high H generation rate, the mass exchange coefficient defined as decreased and the reaction products settled gradually with decreasing the H generation rate. Therefore, the Na concentration decreased at the reaction front and resulted in the SCR terminating naturally.

The Evaluation of the influence of hydride rim and biaxial stress condition on the cladding failure under a reactivity-initiated-accident by using EDC test method

Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

JAEA-Research 2014-025, 34 Pages, 2014/12

JAEA-Research-2014-025.pdf:6.05MB

EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.

Core dynamics analysis of control rod withdrawal test in HTTR (Contract Research)

Takada, Eiji*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Shimakawa, Satoshi; Nojiri, Naoki; Fujimoto, Nozomu

JAERI-Tech 2004-048, 60 Pages, 2004/06

JAERI-Tech-2004-048.pdf:4.18MB

The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395C and coolant outlet temperature of 850C/950C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. The reactivity insertion test demonstrates that rapid increase of reactor power by withdrawing the control rod is restrained by only the negative reactivity feedback effect without operating the reactor power control system, and the temperature transient of the reactor is slow. The best estimated analyses have been conducted to simulate reactor transients during the reactivity insertion test. A one-point core dynamics approximation with one fuel channel model is applied to this analysis. It was found that the analytical model for core dynamics could simulate the reactor power behavior.

Failure thresholds of high burnup BWR fuel rods under RIA conditions

Nakamura, Takehiko*; Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo

Journal of Nuclear Science and Technology, 41(1), p.37 - 43, 2004/01

https://doi.org/10.3327/jnst.41.37

Transient deformation of high burnup BWR fuel rods was measured and failure limit was examined under simulated RIA conditions. Brittle cladding failure occurred at a small strain of about 0.4% during an early phase of the pulse irradiation tests at the NSRR. Strain rates were in an order of tens %/s at the time of the failure. Comparison of the results with thermal expansion of pellets suggested that the deformation was caused by thermal expansion of the pellets. In other words, the influence of fission gases in the pellets was small in the early phase of the deformation. Separate effect tests were conducted to examine influence of the cladding temperature on the failure behavior of cladding. Influence of the pulse width on the failure threshold was discussed in terms of the strain rate, magnitude of the deformation and temperature of the cladding for high burnup BWR fuel rods under RIA conditions.

Behavior of uranium-zirconium hydride fuel under reactivity initiated accident conditions

Sasajima, Hideo; Sugiyama, Tomoyuki; Nakamura, Takehiko; Fuketa, Toyoshi; Uetsuka, Hiroshi

Proceedings of 7th International Topical Meeting on Research Reactor Fuel Management (ENS RRFM2003), p.109 - 113, 2003/03

Uranium-zirconium hydride (U-ZrHx) fuel has been widely utilized in the world as TRIGA reactor fuel. In order to obtain the fuel performance data under accident conditions and to enhance accountability of the safety assessment of the reactors using the fuel, irradiation tests under power burst type accident conditions have been conducted in the NSRR. Five pulse irradiation tests have been performed at peak fuel enthalpies ranging from 187 J/g to 483 J/g. Cladding surface temperature increased rapidly at the pulse and DNB occurred in peak fuel enthalpy over 187 J/g in the tests. The DNB occurred at lower fuel enthalpy in the U-ZrH1.6 fuel than in the UO fuel rods. The rod internal pressure rose up to as high as 1MPa in the transient heating tests, suggesting considerable release of the hydrogen decomposed from the fuel. The peak pressure was lower than equilibrium hydrogen pressure of 1.5MPa at the peak temperature, suggesting the transient effect. Considerable PCMI was observed in the tests, through cladding elongation up to 3.3 mm synchronized to the pellet stack deformation.